Method for preparing high purity 233 uranium

ABSTRACT

A method for preparing high purity 233uranium is provided by processing molten salt reactor fuels containing protactinium isotopes in a series of at least two hold-up tanks wherein in the first tank, decay of 232protactinium to 232uranium is permitted to a preselected value and the 232uranium is removed by subsequent fluorination. In the second tank, decay of 233protactinium to 233uranium is permitted to a preselected value and the 233uranium, which is highly depleted in 232uranium, is recovered by subsequent fluorination.

United States Patent 1 Bell et al.

1 Jan. 30, 1973 METHOD FOR PREPARING HIGH PURITY 233 URANIUM OTHER PUBLICATIONS Jaye et al ..l76/37 [75] Inventors: Michael J. Bell, Marvin E. Whatley, both of Oak Ridgm-rennl AEC document BNL-483, Thorium U233 Symposium, Jan. 9, 1958, pp. 40, 41

[73] Assignee: The United States of America as represented by the United States 'i' Examu}er carl D'Quarfmth Atomic Energy Commission Assistant ExammerF. M. G|ttes, Jr.

Attorney-Roland A. Anderson [22] Filed: June 10, 1970 [57] ABSTRACT [21] Appl' 45086 A method for preparing high purity uranium is provided by processing molten salt reactor fuels contain- [52] U.S.Cl. ..423/5,423/l9, 23/252, ng prota ni m i otopes in a series of at least two 252/3()1 R hold-up tanks wherein in the first tank, decay of 51 Int. Cl. ..C0lg 43/06 232protactinium to uranium is permitted, to a [58] Field of Search ..23/325, 326; 176/37, 49; preselected value and the uranium is removed y 252 30 1 subsequent fluorination. In the second tank, decay of protactinium to "uranium 'is permitted to a [56] References m preselected value and the uranium, which is highly depleted in uranium, is recovered by subsequent UNITED STATES PATENTS fluorination.

3,395,991 8/1968 Grimes ct al. ..23/325 3,046,088 7/l962 Horn ..23/326 3,577,225 5/1971 Shaffer et al. ..23/325 2 1 Draw"; Fgure 3,472,633 l0/l969 McNeeseetal ..23/32S 232 OUAL|TY( t I pp 23% 1o l i A y ,7

O5 5 i if, m

- I o l I:

233 YIELD U 232 6 --X QUALITY (ppm) U 1O PATENTEDJAHO I975 3,714,322

V /F (days) MQQ ATTORNEY.

METHOD FOR PREPARING HIGH PURITY 233 URANIUM BACKGROUND OF THE INVENTION The invention herein was made in the course of, or under, a contract with the United States Atomic Energy Commission. It relates generally to methods for preparing nuclear fuels and more particularly to a method for preparing high purity uranium from molten salt reactor fuels.

The concept of a reactor fueled with a molten metallic salt mixture which contains fissile and/or fertile materials has been widely investigated and has been demonstrated. In the molten salt reactor technology a recent reactor concept which comprised a single fluid, double region molten salt breeder reactor has been fully described in Ser. No. 733,843, filed on June 3, 1968, in the names of Edward S. Bettis et al. for Single Fluid Molten Salt Nuclear Breeder Reactor." Various methods have been devised to reprocess these molten salt fuels. In one method reductive extraction techniques are employed to separate the various contaminants such as protactinium and rare earths from the salt mixture, and in such processes one product stream contains the isotopes of protactinium, specifically protactinium, protactinium and protactini- Previously, the stream containing the protactinium was passed outside of the reactor where the protactinium was permitted to decay to corresponding uranium isotopes which were then recycled to the fuel salt or removed for specific purposes. The uranium normally contained several hundred ppm uranium. If the uranium was to be processed for other than recycle to the reactor, as for example, to produce reactor fuel plates, the activity associated with the uranium and its daughters increased the difficulties of the reprocessing. Accordingly, it is desirable to provide a method for preparing uranium with a very low quantity of' uranium.

SUMMARY OF THE INVENTION The present invention includes a method for preparing high purity uranium. More particularly, the method comprises passing a molten salt reactor fuel mixture containing protactinium isotopes to a first hold-up tank where protactinium is permitted to decay to uranium to a preselected value, fluorinating the molten salt mixture to remove essentially all of the uranium contained therein, passing the remaining mol-, ten salt mixture to a second hold-up tank where protactinium is permitted to decay to uranium to a preselected value, and thereafter fluorinating the molten salt mixture to remove essentially all of the uranium. The product uranium hexafluoride stream from the second fluorinator is highly depleted in uranium and serves as a feed for further processing into reactor fuels. Advantageously, the absence of uranium isotopes in the uranium product greatly simplifies this subsequent processing by reducing the high level of radiation which has heretofore been attendant with the mixed uranium product. A uranium product having a quality (atomic ratio of U/ U) of 1 ppm is readily achieved for molten salt mixtures containing an initial atomic ratio of Pa/ Pa of 50 ppm, using a hold-up time of 8 days in the first tank and a hold-up time of 120 days in the second tank.

BRIEF DESCRIPTION OF THE DRAWING The sole drawing is a graph relating the quality and yield of the "uranium product (as uranium hexafluoride) to the hold-up time for protactinium to uranium decay and the initial Pa/ la concentration ratio.

DESCRIPTION OF THE PREFERRED EMBODIMENT The present invention utilizes as a starting solution a molten salt reactor fuel containing protactinium isotopes. The particular method by which such a solution is provided is not critical to the successful practice of the invention. A convenient source of a molten salt solution containing protactinium isotopes is a product stream from a molten salt reprocessing method. In one of the well known methods for processing molten salt fuels, the molten salt is first fluorinated with fluorine gas to remove the bulk of the uranium and thereafter the fluorinated molten salt is contacted with molten bismuth containing a metal reductant, such as lithium or thorium, in a reductive extraction operation whereby the protactinium values are extracted into the metal phase; the rare earths in a separate extraction unit also are extracted from the molten salt solution into the metal phase. The protactinium-laden metal solution is next passed to an oxidizer where in the presence of hydrogen fluoride and a side stream of fluorinated molten salt solution the extracted protactinium metal values are oxidized to protactinium tetrafluoride and extracted into the salt phase. The exiting salt solution from the oxidizer has a protactinium content, which is concentrated by a factor of at least 50 times the reactor concentration and serves as the product feed to the hold-up tanks.

In accordance with the method of this invention the molten salt solution containing the protactinium values is passed to a first hold-up tank wherein the decay of protactinium to uranium is permitted to a preselected value. The flow through this hold-up tank should approximate plug flow, that is, no mixing, so that each portion of the salt is detained the same length of time. In practice, this may be effected by using, for example, a heated serpentine tube of proper length. The molten salt solution is then fluorinated with fluorine gas at 525 C. to remove essentially all of the uranium values, which comprise an impure mixture of uranium and uranium, as uranium hexafluoride. The protactinium values, not forming a volatile fluoride, remain in the molten salt solution. Thereafter the fluorinated salt solution is passed to a second holdup tank of the same .type wherein the decay of protactinium to uranium is permitted to a preselected value. Finally, the molten salt solution is again fluorinated with fluorine gas at' 525 C. to remove es-v sentially all of the uranium as a purified productl uranium hexafluoride). Again essentially no protactinium is removed in the fluorination.

The quality of the product, defined as the atomic ratio of uranium to uranium, in the product is given by where A, is the radioactive disintegration constant for protactinium and equals 0.525 da"; A, is the radioactive disintegration constant for protactinium and equals 0.0253 da, V and V are the volumes of the first and second hold-up tanks, respectively, and F is the volumetric flow rate of the molten salt solution. Referring to the drawing, the transient time in the first hold-up tank is selected for the desired quality of the product, knowing the initial Pa/Wa ratio. The quality for such a system for initial atomic ratios of Pa/ Pa of 50, 10, and 5 ppm are given in the drawing for a transient time (V /F) equal to 120 days as a function of the transient time (V,/F) in the first hold-up tank. Although a transient time of 80 to l days for the second hold-up tank is preferred, it will be apparent here that other transient times in the second hold-up tank may be used provided they are sufficiently long in comparison to the transient times for the first hold-up tank. Where a time longer than 120 days is employed, there is little effect upon the yield or the quality of the product.

While the transient time in the first hold-up tank will vary depending upon the quality of the product and the initial atomic ratio of Pal -Pa, it is preferred that the transient time should be from 8 to 12 days. Longer decay times reduce the quantity of U in the product but also reduce the product yield. Longer times in either hold-up tank increase tank construction costs. Under such process conditions the quality of the product for initial atomic ratios of Pa/ Pa of 50, 10, and 5 ppm, is l, 0.2, and 0.1 ppm, respectively.

The yield of the system, i.e., the fraction of protactinium recovered as uranium is given by and for the second fluorinator is While the present invention provides for the production of high purity uranium using molten salt reactor fuels which contain protactinium isotopes, it also affords a convenient method for disposing of "protactinium isotopes from these molten salt solutions. Because ofits long half-life, essentially all protactinium remains in the molten salt solution. The protactinium isotopes may then be disposed of by either discarding the salt solutionor removing it from the salt solution, such as by reductive extraction, prior to recycle of the salt solution to the reactor.

It will be apparent to those skilled in the art that elevated temperatures are required for the successful operation of the present invention. The molten salt reactor fuel solution should be maintained at a temperature above the liquidus temperature of the particular molten salt mixture employed. To insure a safe margin for process control, it is preferred that the temperature of the salt solution be maintained at least 25 C. above the liquidus temperature of the molten salt mixture and temperatures above about 800 C. are not recommended. The liquidus for the carrier salt mixture LiF BeF ThF, (72 l6 12 mole percent) is 500 C. The fluorinators should be operated at a temperature of about 525C.

What is claimed is:

l. A method for isolating uranium isotopes from a single fluid molten salt reactor fuel solution containing protactinium isotopes comprising the steps of passing said solution to a first hold-up tank, holding up said salt solution for about 8 to 12 days for *protactinium to decay to uranium to a preselected value, said uranium being present as uranium tetrafluoride, fluorinating the exiting salt solution from said first hold-up tank at a temperature of at least 25 C. above the liquidus temperature of said solution so as to form uranium hexafluoride from said uranium tetrafluoride, thus removing essentially all of the uranium in said solution as uranium hexafluoride, passing the fluorinated salt solution to a second hold'up tank, holding up said fluorinated salt solution for about to days for protactinium to decay to uranium to a preselected value and thereafter fluorinating the exiting salt solution from said second hold-up tank at a temperature of at least 25 C. above the liquidus temperature of said solution thus removing essentially all of the uranium formed from said decay of protactinium as uranium hexafluoride.

2. The method of claim 1 wherein said molten salt reactor fuel comprises LiF BeF ThF, UF, (71.7 l 6 12 0.3 mole percent) having a total protactinium concentration of 0.003 mole percent and the quality of the uranium hexafluoride is 1 ppm U in U. 

1. A method fOr isolating 233uranium isotopes from a single fluid molten salt reactor fuel solution containing protactinium isotopes comprising the steps of passing said solution to a first hold-up tank, holding up said salt solution for about 8 to 12 days for 232protactinium to decay to 232uranium to a preselected value, said 232uranium being present as uranium tetrafluoride, fluorinating the exiting salt solution from said first hold-up tank at a temperature of at least 25* C. above the liquidus temperature of said solution so as to form uranium hexafluoride from said uranium tetrafluoride, thus removing essentially all of the uranium in said solution as uranium hexafluoride, passing the fluorinated salt solution to a second hold-up tank, holding up said fluorinated salt solution for about 80 to 120 days for 233protactinium to decay to 233uranium to a preselected value and thereafter fluorinating the exiting salt solution from said second hold-up tank at a temperature of at least 25* C. above the liquidus temperature of said solution thus removing essentially all of the uranium formed from said decay of 233protactinium as 233uranium hexafluoride. 